Very High Temperature Gas-cooled Reactor Core Research Articles

Preliminary neutronic analyses for annular VHTR with different coated fuel particle designs

Abstract The purpose of the present work is to determine the neutronic characteristics of annular cylindrical prismatic Very High Temperature Gas-cooled Reactor (VHTR) with different advanced fuel designs. The study includes detailed neutronic investigations of the eight types of TRISO (containing SiC layer) and TRIZO (containing ZrC layer) fuel particles for annular VHTR core. All calculations were performed using continuous energy Monte Carlo code MVP2.0 and MVPBURN with JENDL4.0 nuclear data library. The obtained results indicated that SiC + PyC and PyC layers in TRISO fuel appeared to be not excellent choice for the annular VHTR core due to its shorter operation year and, lower fuel burnup. The annular cylindrical prismatic core fueled with TRIZO fuel particle shows good neutronic characteristics compared to those designed with other TRISO fuel particles.

Time-dependent neutron diffusion analysis using finite element method for a block-type VHTR core design

The very high-temperature gas-cooled reactor (VHTR) has attracted interest owing to its high passive safety. A reactor analysis tool is required to analyze the safety characteristics of a VHTR in detail, and such a tool should be capable of a transient calculation for a reactivity insertion accident such as control rod ejection or withdrawal. This work presents a transient analysis of a block-type VHTR core. Time-dependent neutron diffusion equation is solved by the finite element method. A simplified thermal fluid analysis tool is also implemented to consider thermal feedback. In addition, a new method is introduced to resolve the control rod cusping effect without additional computation or mesh reconstruction. The above methods are applied to a reactor physics code CAPP, developed at Korea Atomic Energy Research Institute (KAERI). The methods are tested on three-dimensional VHTR problems that include control rods and show encouraging results.

VHTR core analysis with McCARD and DeCART codes for high temperature engineering test reactor benchmark

This paper presents a Very High Temperature Gas-Cooled Reactor (VHTR) core analysis of the McCARD and DeCART-2D with High Temperature Engineering Test Reactor (HTTR) benchmark problem. The numerical results of the HTTR benchmark problem are demonstrated, the capability for VHTR core analysis of McCARD is validated by using the experiment data, and that of DeCART-2D is verified by comparing the numerical results with those of McCARD. The detail HTTR benchmark problem and specification of core geometry and material are also described. Also, a method of approximation for the complex geometry of the VHTR core is described for modelling in computational code. In terms of the multiplication factor and temperature coefficient, the numerical results of McCARD are easily compared with the HTTR experiment data. Because the DeCART-2D can solve 2-dimensional (2D) geometry, the simulation of 2D HTTR fuel rod, block, and core problem is performed with DeCART-2D and McCARD. The numerical results between them are around 200–400 pcm, meaning that DeCART-2D can be used to analyze VHTR core.

Numerical solution of heat transfer process in a prismatic VHTR core accompanying bypass and cross flows

Bypass flow and cross flow gaps both exist in the core of a very high temperature gas-cooled reactor (VHTR), which is inevitable owing to tolerances in manufacturing, thermal expansion and irradiation shrinkage. The coolant mass flow rate distribution, temperature distribution, and hot spot temperature are significantly affected by bypass and cross flows. In the present study, three-dimensional CFD analysis is conducted for thermal analysis of the reactor core. A validation study for the turbulence model is performed by comparing the friction coefficient with published correlations. A sensitivity study of the near wall mesh is conducted to ensure mesh quality. Parametric studies are performed by changing the size of the bypass and cross gaps using a one-twelfth sector of a fuel block. Simulation results show the influence of the bypass gap size on temperature distribution and coolant mass flow rate distribution in the prismatic core. It is shown that the maximum fuel and coolant channel outlet temperatures increase with an increase in the gap size, which may lead to a structural risk to the fuel block. The cross flow is divided into two types: the cross flow from the bypass gap to the coolant channels and the cross flow from the high-pressure coolant channels to low-pressure coolant channels. These two types of flow have an opposing influence on the temperature gradient. It is found that the presence of the cross flow gaps may have a significant effect on the distribution of the coolant in the core due to flow mixing in the cross gaps.

Analytical Study on the Effective Thermal Conductivity of VHTR Fuel Block Geometry with Multiple Cylindrical Holes

In Korea, the Very High Temperature Gas-Cooled Reactor (VHTR) PMR200 is being developed in the Nuclear Hydrogen Development and Demonstration project. Its core consists of hexagonal prism-shaped graphite blocks for the fuel and reflector, and each hexagonal fuel block contains 108 cylindrical coolant holes and 210 fuel compacts. Because of these holes and fuels, the heat transfer in lateral directions in the fuel blocks becomes very complicated. Especially in accident situations when forced convection is lost, the majority of the afterheat flows in the radial direction by conduction across the large number of coolant holes. Moreover, radiation heat transfer is supposed to be added to the radial heat transfer modes owing to the high temperature of the VHTR core. Because of these complexities in radial heat transfer, reliable modeling for effective thermal conductivity (ETC) is required in order to analyze the reactor core thermal behavior using lumped-parameter codes, which are often used to evaluate the integrity of nuclear fuel embedded in the graphite block. In this study, the ETC model adopted in the GAMMA+ code was introduced, and the adequacy of the model was assessed by the commercial computational fluid dynamics (CFD) code CFX-13. The results of the CFD analysis were consistent with the ETC model in general even if a slight disagreement was shown for the case of high temperature. From these analyses, it could be concluded that the ETC model adopted in the GAMMA+ code is an adequate model for the analysis of the PMR200 reactor core. Moreover, it was found that the effect of fuel gap can cause an overprediction of the ETC if the fuel compact thermal conductivity is larger than the applicable range of the model.

CFD Analysis and Assessment for Cross-Flow Phenomena in VHTR Prismatic Core

The very high temperature gas-cooled reactor (VHTR) is a uranium-fueled, graphite-moderated, and helium-cooled reactor envisioned as one of the promising future nuclear reactor concepts due to its high efficiency, safety, and variety of applications. In this reactor concept, core bypass flow and cross flow are currently considered as key issues because of their inherent uncertainties and complication of prediction. Recently, the computational fluid dynamics (CFD) method has received a great deal of attention as a method for understanding the flow behavior in the VHTR core, including the bypass and the cross flow. However, validation of this method has not been sufficiently conducted yet based on real experimental data. For this reason, prediction capability of the CFD method was validated in this study by comparing the predictions with the existing multi-hole experimental data obtained by Groehn as a part of the core thermal-hydraulics design study for the NHDD PMR-200. A two-stacked fuel block with wedge-shaped cross gap was simulated for computational domain as the experimental setup. The flow loss coefficients and the velocity distributions of the cross flow from the experiment were compared to the CFD predictions extensively. As a result, good agreements between the CFD predictions and the experimental results were observed, confirming prediction capability of the CFD method for the complicated cross flow in the VHTR core. Furthermore, the velocity distributions and pressure distribution in the cross gap were investigated to identify the characteristics of the cross flow.

フランジ型燃料ブロックを用いた高温ガス炉の炉心有効流量

The objective of present study is to evaluate and to increase the effective coolant flow rate of VHTR (Very High Temperature gas-cooled Reactor) core whose thermal output is 50MW and core outlet gas temperature is 950°C, without changing the sizes of the core and fuel element, Four types of fuel element are discussed, they are two flange type pin-in-block fuel elements which have 18 fuel rods (P 18) and 36 fuel rods (P 36), a flange type multihole fuel element (MH 8) and a dowel type pin-in-block fuel element (P 15).The flow distribution analysis was carried out by flow network code which calculated coolant pressure, flow rate and temperature. As the result, the followings were obtained:Effective coolant flow rate (Weff) became lower in the order, MH 8, P 36, P 18 and P 15 fuels. Weff of the P 15 fuel was the lowest, because crossflow loss coefficient of dowel type fuel element was lower than that of flange type. Weff of the P 36 fuel was higher than that of the P 18 fuel, because of larger total cross section of coolant flow channels. Weff of the MH 8 fuel was higher than that of the P 36 fuel, because of larger equivalent diameter of the channels.This analysis clearly revealed that the effective coolant flow rate increased by using the fuel element with large values of crossflow loss coefficient, cross section and equivalent diameter of coolant flow channels.

Experimental studies on the thermal and hydraulic performance of the fuel stack of the VHTR: Part II. HENDEL multi-channel test rig with twelve fuel rods

Tests on heat transfer and fluid-dynamics of the mock-up fuel stack of the Very High Temperature Gas Cooled Reactor (VHTR) were performed by the multi-channel test rig (T 1 - M) of the Helium Engineering Demonstration Loop (HENDEL). The T 1 - M simulated one column of the fuel stack in the VHTR core and contained twelve simulated fuel rods. The heat generation rate of each fuel rod was varied to simulate the power distribution of the VHTR core in the horizontal plane. In parallel with this experiment, a three-dimensional thermal analysis code was developed in order to check the experimental temperature distribution of the fuel stack. Experimental results showed that the distribution of the helium gas flow rate was influenced by the temperature distortion in the mock-up fuel stack. The maximum deviation of the helium gas flow rate from the mean value was 10% in the case of an asymmetric power distribution test at a low Reynolds number. The variation of the calculated temperature distribution in the fuel stack was about 17-35°C in each case, indicating that the temperature distortion in the fuel stack was flattened by thermal conduction in the graphite block.

Experimental studies on the thermal and hydraulic performance of the fuel stack of the VHTR: Part I. HENDEL single-channel tests with uniform heat flux

The Helium Engineering Demonstration Loop (HENDEL) was constructed for a large-scale component test of the very high temperature gas-cooled reactor (VHTR) under simulated reactor operating conditions. The fuel stack test section (T 1) of HENDEL simulates the fuel stack of the VHTR core to investigate thermal and hydraulic performance. Experimental results showed that the turbulent heat transfer characteristics were maintained down to a low Reynolds number of about 2000. Friction factors and heat transfer coefficients for the fuel rod were found to be higher than those in a concentric smooth annular channel.

高温ガス炉フランジ型燃料ブロックのクロス流れ

Fuel element design incorporating mating flanges at block end faces has the potential to improve thermal hydraulic performance of a VHTR (very high temperature gas-cooled reactor) core. As part of research and development efforts to establish flange type fuel element design, experiments and analyses were carried out on crossflow through interface gap between elements. Air at atmospheric pressure and ambient temperature was used as a fluid. Crossflow loss coefficient factors were obtained with three test models, having different flange mating clearances, for various interface gap configurations, gap widths and block misalignments. It was found that crossflow loss coefficient factors for flange type fuel element were much larger than those for conventional flat-faced element. Numerical analyses were also made using a simple model devised to represent the crossflow path at the fuel element interface. The close agreement between numerical results and experimental data indicated that this model could predict well the crossflow characteristics of the flange type fuel element.

高温ガス実験炉燃料体の伝熱流動試験 ＩＩＩ ＨＥＮＤＥＬ多チャンネル試験装置による不均一発熱試験結果と解析

Experimental studies on heat transfer and fluid-dynamic characteristics of a mock-up fuel stack of the Very High Temperature Gas Cooled Reactor (VHTR) were carried out by the use of a multi-channel test rig of Helium Engineering Demonstration Loop (HENDEL). Heat generation rate of each simulated fuel rod was varied independently to simulate a power distribution of the VHTR core in a horizontal plane. In parallel with this experiment, a three-dimensional thermal analysis code was developed in order to check the experimental temperature distribution of the fuel stack.Experimental results showed that the distribution of coolant flow rate varied with the temperature distortion in the fuel stack. The variation of calculated temperature distribution in the fuel stack was about 20-35°C in each case, indicating that the temperature distortion in the fuel stack was flattened by thermal conduction through the graphite.

多目的高温ガス実験炉炉心のクロス流れ ＩＩ ２ブロッククロス流れの解析

In the first paper, two-block crossflow experiments were conducted to develop crossflow coefficient carrelatians, since crossflow, which is inherent in the VHTR(very high-temperature gas-cooled reactor) core, has a significant effect on the core flow distribution.This paper proposes a numerical model with which the global crossflow loss coefficient factors for the parallel and wedge-shaped gaps are calculated. The fuel element crossflow domain was discretized by triangular elements and the pressure distribution in the block interface gap was calculated by the finite element method. The crossflow rate from the periphery of the block to the coolant channels was evaluated from the pressure distribution. The numerical results agreed well with the experimental results. In addition, empirical crossflow equations were devised for the analysis of flow distribution in the VHTR core.

多目的高温ガス実験炉炉心のクロス流れ Ｉ ２ブロッククロス流れ実験

In course of the design development of the VHTR (very high-temperature gas-cooled reactor) core, special attention has been given to the crossflow through the interface gaps between fuel blocks because of its adverse effect on thermal hydraulic performance of the core.Experiments were carried out to determine crossflow rate on full-scale two-block models based on a proposed design of fuel element. Air at atmospheric pressure and ambient temperature was used as a fluid. Crossflow rate and pressure difference were measured for various gap configurations and widths. Measurements of the crossflow rate that entered into individual coolant channel were also made.The experimental results were expressed by the relation between crossflow loss coefficient factor and Reynolds number with interface gap as a parameter, allowing the prediction of crossflow in the core thermal hydraulic analysis.